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学者姓名:吴宏春

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< Page ,Total 45 >
NECP-Atlas: A new nuclear data processing code EI Scopus SCIE
期刊论文 | 2019 , 123 , 153-161 | Annals of Nuclear Energy
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Abstract :

A new nuclear data processing code called NECP-Atlas is under development at Xi'an Jiaotong University in China. The motivation for the development is to establish a platform to carry out deeper researches on nuclear data processing methods to satisfy the demands on accurate cross sections in the fields of high-fidelity transport simulation and advanced reactor design. The present paper describes the methods used in the current version of NECP-Atlas and demonstrates the performance and accuracy of the code on a variety of benchmarks. At the present time, NECP-Atlas can process ENDF/B-VII.1, ENDF/B-VII.0, CENDL-3.1, JEFF-3.2 and JENDL-4.0 evaluations, and generate WIMS-D and ACE format libraries. The accuracy of the code is comparable with NJOY2016. NECP-Atlas is competent to provide cross sections for deterministic and Monte Carlo transport calculations. © 2018 Elsevier Ltd

Keyword :

Advanced reactor design Cross section Multi band NECP-Atlas Resonance elastic scattering Transport calculation Transport simulation Xi'an Jiaotong University

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GB/T 7714 Zu, Tiejun , Xu, Jialong , Tang, Yongqiang et al. NECP-Atlas: A new nuclear data processing code [J]. | Annals of Nuclear Energy , 2019 , 123 : 153-161 .
MLA Zu, Tiejun et al. "NECP-Atlas: A new nuclear data processing code" . | Annals of Nuclear Energy 123 (2019) : 153-161 .
APA Zu, Tiejun , Xu, Jialong , Tang, Yongqiang , Bi, Huchao , Zhao, Fei , Cao, Liangzhi et al. NECP-Atlas: A new nuclear data processing code . | Annals of Nuclear Energy , 2019 , 123 , 153-161 .
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An improved fitting method for subgroup parameters based on the heterogeneous cells EI SCIE
期刊论文 | 2019 , 56 (2) , 179-192 | Journal of Nuclear Science and Technology
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Abstract :

An accurate subgroup parameters fitting method, where background cross sections obtained based on heterogeneous cells are used to fit the subgroup level and subgroup weight, is proposed in this paper. Due to the dependence of background cross section on the subgroup level, the calculation of the subgroup parameters is a nonlinear problem, which causes the iteration between fitting subgroup parameters and updating background cross sections. The cubic spline interpolation method is used to update the background cross sections to avoid frequently solving fixed source equations. In the fitting process, the negative subgroup parameters are often obtained, and the accuracy of the subgroup parameters is very sensitive to the iterative initial values of subgroup levels. To avoid these problems, additional constraints ensuring positive subgroup parameters and guaranteeing numerical stability are added to the optimization function. Penalty function method is used to convert the optimization problem with constraints into the one without constraints, making the problem easy to be solved. The proposed method is tested against the problems of pin cell, pressurized water reactor assemblies and plate-type assembly. The numerical results show that the self-shielded cross sections calculated by the proposed method agree well with those by Monte Carlo code. © 2018, © 2018 Atomic Energy Society of Japan. All rights reserved.

Keyword :

Cubic spline interpolation method Fitting method Nonlinear problems Optimization function Optimization problems Penalty function methods Subgroup method subgroup parameters

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GB/T 7714 Zu, Tiejun , Xia, Fan , Wu, Hongchun . An improved fitting method for subgroup parameters based on the heterogeneous cells [J]. | Journal of Nuclear Science and Technology , 2019 , 56 (2) : 179-192 .
MLA Zu, Tiejun et al. "An improved fitting method for subgroup parameters based on the heterogeneous cells" . | Journal of Nuclear Science and Technology 56 . 2 (2019) : 179-192 .
APA Zu, Tiejun , Xia, Fan , Wu, Hongchun . An improved fitting method for subgroup parameters based on the heterogeneous cells . | Journal of Nuclear Science and Technology , 2019 , 56 (2) , 179-192 .
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Heterogeneous discontinuity factor treatment in Variational Nodal Method EI
期刊论文 | 2019 , 127 , 341-350 | Annals of Nuclear Energy
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Abstract :

To handle the control rod cusping effect in Pressurized Water Reactor (PWR) core calculation, the heterogeneous Variational Nodal Method (VNM) employed by the fuel management calculation code system NECP-Bamboo has been enhanced to treat heterogeneous discontinuity factor (DF) appearing on nodal interface. To solve the neutron-diffusion equations with heterogeneous DF, firstly, the functional in VNM is modified to contain the discontinuity of neutron flux in the surface integral term. Secondly, other than volumetric flux and surface partial currents, cross sections and surface DF are also expanded into the sum of orthogonal piece-wise polynomials to construct the nodal response matrixes. Four test problems in this paper including the CISE, Henry-Worley, HAFAS benchmark problems and the BEAVRS problems were employed to verify the method in treating heterogeneous DF. It has been demonstrated that the control rod cusping effect can be fully eliminated by the heterogeneous VNM with heterogeneous DF in terms of control rod differential worth and three-dimensional pin-power profile. © 2018 Elsevier Ltd

Keyword :

Bench-mark problems Core calculations Differential worth Discontinuity factor Neutron diffusion equations Partial currents Surface integrals Variational nodal method

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GB/T 7714 Li, Yunzhao , Liang, Boning , Wu, Hongchun et al. Heterogeneous discontinuity factor treatment in Variational Nodal Method [J]. | Annals of Nuclear Energy , 2019 , 127 : 341-350 .
MLA Li, Yunzhao et al. "Heterogeneous discontinuity factor treatment in Variational Nodal Method" . | Annals of Nuclear Energy 127 (2019) : 341-350 .
APA Li, Yunzhao , Liang, Boning , Wu, Hongchun , Li, Zhipeng , Yang, Jiewei . Heterogeneous discontinuity factor treatment in Variational Nodal Method . | Annals of Nuclear Energy , 2019 , 127 , 341-350 .
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Validation of SARAX for the China Fast Reactor with the extrapolated experimental data EI
期刊论文 | 2019 , 127 , 188-195 | Annals of Nuclear Energy
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Abstract :

The validation works have been implemented to a newly-developed code SARAX for China Fast Reactor (CFR) in this paper. Different with the conventional way to validate the code using the dedicated experimental data, a theoretical approach was proposed and implemented by using the existing similar experimental data. This theoretical approach is based on the technology of sensitivity/uncertainty analysis, similarity analysis and nuclear-data adjustment. In our previous works, detailed introduction towards sensitivity/uncertainty analysis and similarity analysis have been implemented to distinguish the existing experiments which are 'similar’ to the CFR's criticality characteristics. This paper focus on the method to extrapolate the 'similar’ experimental data to predict the best-estimate measurements of CFR with application of nuclear-data adjustment, providing reference values to validate the SARAX code. From the numerical results, it can be observed that through nuclear-data adjustment, the simulation results of the existing experiments can agree well with corresponding measurements, with the bias reduced notably to be within 25 pcm. Moreover, the nuclear-data adjustment can also improve the nuclear-data uncertainties notably, with the relative uncertainties of the reactor keff due to nuclear data having been reduced from the value exceeding 1.0% to the values about 0.15%. © 2018

Keyword :

Code validation Nuclear data Numerical results Reference values Relative uncertainty Sensitivity/uncertainty analysis Similarity analysis Theoretical approach

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GB/T 7714 Wan, Chenghui , Qiao, Liang , Zheng, Youqi et al. Validation of SARAX for the China Fast Reactor with the extrapolated experimental data [J]. | Annals of Nuclear Energy , 2019 , 127 : 188-195 .
MLA Wan, Chenghui et al. "Validation of SARAX for the China Fast Reactor with the extrapolated experimental data" . | Annals of Nuclear Energy 127 (2019) : 188-195 .
APA Wan, Chenghui , Qiao, Liang , Zheng, Youqi , Cao, Liangzhi , Wu, Hongchun . Validation of SARAX for the China Fast Reactor with the extrapolated experimental data . | Annals of Nuclear Energy , 2019 , 127 , 188-195 .
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The material-region-based 2D/1D transport method EI
期刊论文 | 2019 , 128 , 1-11 | Annals of Nuclear Energy
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Abstract :

The 2D/1D transport method is the dominant method for high-fidelity direct whole-core transport calculations, which attracts a lot of attention in recent decades. In the 2D/1D method, some sources of deviation are introduced, including spatial and angular approximation of leakage term and cross-section homogenization for 1D axial calculation. These approximations are analyzed and a material-region-based 2D/1D transport method with anisotropic leakage term, which avoids cross-section homogenization, as well as reduces spatially flat leakage approximation, are developed to improve accuracy at the expense of acceptable memory and efficiency loss. Finally, a BWR assembly case, the C5G7 benchmark and a rod-cluster assembly case are tested to verify the accuracy and performance of the material-region-based 2D/1D method. © 2018 Elsevier Ltd

Keyword :

Cluster assembly Efficiency loss High-fidelity Leakage terms Material region NECP-X Transport method Whole-core transport

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GB/T 7714 Liu, Zhouyu , Zhao, Chen , Cao, Lu et al. The material-region-based 2D/1D transport method [J]. | Annals of Nuclear Energy , 2019 , 128 : 1-11 .
MLA Liu, Zhouyu et al. "The material-region-based 2D/1D transport method" . | Annals of Nuclear Energy 128 (2019) : 1-11 .
APA Liu, Zhouyu , Zhao, Chen , Cao, Lu , Wu, Hongchun , Cao, Liangzhi . The material-region-based 2D/1D transport method . | Annals of Nuclear Energy , 2019 , 128 , 1-11 .
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The pseudo-resonant-nuclide subgroup method based global-local self-shielding calculation scheme EI SCIE Scopus
期刊论文 | 2018 , 55 (2) , 217-228 | JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
WoS CC Cited Count: 4
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Abstract :

The pseudo-resonant-nuclide subgroup method (PRNSM) based global-local self-shielding calculation scheme is proposed to simultaneously resolve the local self-shielding effects (including spatial self-shielding effect and the resonance interference effect) for large-scale problems in reactor physics calculations. This method splits self-shielding calculation into global calculations and local calculations. The global calculations obtain the Dancoff correction factor for each pin cell by neutron current method. Then an equivalent one-dimensional (1D) cylindrical problem for each pin cell is isolated from the lattice system by preserving Dancoff correction factor. The local calculation is to perform self-shielding calculations of the equivalent 1D cylindrical problem by the PRNSM. The numerical results show that PRNSM obtains accurate spatial dependent self-shielded cross sections and improves the accuracy of dealing with the resonance interference over the conventional Bondarenko iteration method and the resonance interference factor method. Furthermore, because both global and local calculation is linearly proportional to the size of problems, the global-local calculation scheme could be applied to large-scale problems.

Keyword :

numerical simulation resonance interference reactor physics spatial self-shielding Resonance pseudo resonant nuclide

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GB/T 7714 Liu, Zhouyu , He, Qingming , Zu, Tiejun et al. The pseudo-resonant-nuclide subgroup method based global-local self-shielding calculation scheme [J]. | JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY , 2018 , 55 (2) : 217-228 .
MLA Liu, Zhouyu et al. "The pseudo-resonant-nuclide subgroup method based global-local self-shielding calculation scheme" . | JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 . 2 (2018) : 217-228 .
APA Liu, Zhouyu , He, Qingming , Zu, Tiejun , Cao, Liangzhi , Wu, Hongchun , Zhang, Qian . The pseudo-resonant-nuclide subgroup method based global-local self-shielding calculation scheme . | JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY , 2018 , 55 (2) , 217-228 .
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Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios EI SCIE Scopus
期刊论文 | 2018 , 111 , 271-279 | ANNALS OF NUCLEAR ENERGY
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Abstract :

A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order. (C) 2017 Elsevier Ltd. All rights reserved.

Keyword :

Accelerator Driven System Pyro-chemical reprocessing Minor actinide transmutation Flexibility analyses

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GB/T 7714 Zhou, Shengcheng , Wu, Hongchun , Zheng, Youqi . Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios [J]. | ANNALS OF NUCLEAR ENERGY , 2018 , 111 : 271-279 .
MLA Zhou, Shengcheng et al. "Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios" . | ANNALS OF NUCLEAR ENERGY 111 (2018) : 271-279 .
APA Zhou, Shengcheng , Wu, Hongchun , Zheng, Youqi . Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios . | ANNALS OF NUCLEAR ENERGY , 2018 , 111 , 271-279 .
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Predicting spatially dependent reaction rate for problem with nonuniform temperature distribution by subgroup method EI SCIE Scopus
期刊论文 | 2018 , 111 , 188-203 | ANNALS OF NUCLEAR ENERGY
WoS CC Cited Count: 2 SCOPUS Cited Count: 2
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Abstract :

The subgroup methods based on partial cross section fit scheme (PXSFS) and simplified partial cross section fit scheme (SPXSFS) are proposed in this paper to treat problems with non-uniform temperature distribution. These methods fit the cross sections at different temperatures as partial cross sections and share a same set of subgroup probabilities. The new methods are compared to the pre-existing methods: conventional subgroup method (CSM), the correlation model (CM), the subgroup level adjustment scheme (SLAS) and the number density adjustment scheme (NDAS). The numerical results show that the new methods can better predict the spatially dependent reaction rates than pre-existing methods. Within the new methods, the simplified scheme consumes less computation time and is more numerically stable. Additionally, the superhomogenization (SPH) correction method is studied, which is used to treat the multi-group (MG) equivalence effect. It is found that the subgroup-one-group (subgroup-1G) calculation can fully capture the MG equivalence effect. (C) 2017 Elsevier Ltd. All rights reserved.

Keyword :

Non-uniform temperature distribution Subgroup method Multi-group equivalence effect SPH correction

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GB/T 7714 He, Qingming , Cao, Liangzhi , Wu, Hongchun et al. Predicting spatially dependent reaction rate for problem with nonuniform temperature distribution by subgroup method [J]. | ANNALS OF NUCLEAR ENERGY , 2018 , 111 : 188-203 .
MLA He, Qingming et al. "Predicting spatially dependent reaction rate for problem with nonuniform temperature distribution by subgroup method" . | ANNALS OF NUCLEAR ENERGY 111 (2018) : 188-203 .
APA He, Qingming , Cao, Liangzhi , Wu, Hongchun , Forget, Benoit , Smith, Kord . Predicting spatially dependent reaction rate for problem with nonuniform temperature distribution by subgroup method . | ANNALS OF NUCLEAR ENERGY , 2018 , 111 , 188-203 .
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Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code EI SCIE Scopus
期刊论文 | 2018 , 42 (1) , 261-275 | INTERNATIONAL JOURNAL OF ENERGY RESEARCH
SCOPUS Cited Count: 1
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Abstract :

A molten salt reactor (MSR) is characterized by simultaneously using liquid fuel salt as both the nuclear fuel and coolant. The redistribution of delayed neutron precursors (DNPs) makes the transient behavior of MSRs different from traditional solid-fuel reactors. In this study, a 3D coupled neutronics/thermal hydraulics code, MOREL2.0, was employed to analyze a liquid-fuel Thorium Molten Salt Reactor (TMSR-LF) under perturbations of fuel pump start-up and coast-down and by overheating and overcooling the inlet fuel temperature. Some transient processes were simulated to provide guidance for the future design and optimization of TMSR-LFs. In response to the perturbations, reactivity was lost and gained in the pump start-up and coast-down, respectively. Overheating the inlet fuel temperature introduced negative reactivity, and TMSR-LF maintained a safety state, while overcooling the inlet fuel temperature resulted in positive reactivity. Overcooling by 70 K produced a supercritical transient condition and a rapid increase in power within a short period, which was followed by a decrease in power due to negative temperature feedback. The transient results demonstrate that the negative temperature feedback coefficients guarantee TMSR-LF inherent safety and the variation range of temperature stay within the safety margin.

Keyword :

TMSR transient thermal-hydraulics neutronics molten salt reactor

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GB/T 7714 Cao, Liangzhi , Zhuang, Kun , Zheng, Youqi et al. Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code [J]. | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2018 , 42 (1) : 261-275 .
MLA Cao, Liangzhi et al. "Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code" . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH 42 . 1 (2018) : 261-275 .
APA Cao, Liangzhi , Zhuang, Kun , Zheng, Youqi , Hu, Tianliang , Wu, Hongchun . Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2018 , 42 (1) , 261-275 .
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Improved leakage splitting method for the 2D/1D transport calculation EI SCIE Scopus
期刊论文 | 2018 , 105 , 202-210 | PROGRESS IN NUCLEAR ENERGY
SCOPUS Cited Count: 1
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Abstract :

The 2D/1D transport method is widely implemented for direct whole-core calculations because of its good balance between efficiency and accuracy. In the 2D/1D transport method, negative total source may occur for MOC calculation in iteration process, which will lead to iteration divergence. The negative total source in the MOC equation is studied and an improved leakage splitting method is proposed to improve stability for 2D/1D transport method in this paper. With the new method, the accuracy is preserved while the memory cost only increases slightly. Finally, high leakage cases, the C5G7 and VERA benchmark are tested to show the accuracy and performance of the proposed method.

Keyword :

Improved leakage splitting method NECP-X 2D/1D transport method Direct transport calculation

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GB/T 7714 Zhao, Chen , Liu, Zhouyu , Liang, Liang et al. Improved leakage splitting method for the 2D/1D transport calculation [J]. | PROGRESS IN NUCLEAR ENERGY , 2018 , 105 : 202-210 .
MLA Zhao, Chen et al. "Improved leakage splitting method for the 2D/1D transport calculation" . | PROGRESS IN NUCLEAR ENERGY 105 (2018) : 202-210 .
APA Zhao, Chen , Liu, Zhouyu , Liang, Liang , Chen, Jun , Cao, Liangzhi , Wu, Hongchun . Improved leakage splitting method for the 2D/1D transport calculation . | PROGRESS IN NUCLEAR ENERGY , 2018 , 105 , 202-210 .
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